Modular High-Temperature Gas-Cooled Reactor (HTR) could fulfill the
safety goals of Gen-IV nuclear reactors and its fuel characteristics
provide high confidence in the practical elimination of large
radioactive release from nuclear power plants. Following the concept of
design to safety, the first demonstration project of modular HTR in the
world (HTR-PM, High-Temperature Gas-Cooled Reactor-Pebble-bed Modules)
is under construction in Shidao Bay (in Shandong, China) and is planned
to operate at the end of 2019 [1, 2]. Although HTR-PM is a kind of
advanced reactor with inherent safety, the assessment of radioactivity
during the operation, which plays an essential role in the radiation
protection, is still important [3, 4]. Source term analysis may supply a
radiation safety analysis basis for HTR and could provide the
generation, quantity, release, and radiation hazard of radionuclides in
a nuclear power plant under Normal and accident conditions, making the
design and safety assessment solid and credible.

With the development of nuclear reactor technology, many source
term analysis codes have been developed and broad used. KORIGEN, which
is developed at FZK on the basis of the Oak Ridge Isotope Generation and
Depletion code ORIGEN is used for radionuclides inventory estimation in
the reactor core [5]. The French Institut de Radioprotection et de
Surete Nucleaire (IRSN) and the German Gesellschaft fur Anlagen und
Reaktorsicherheit mbH (GRS) have developed a system of calculation
codes, Accident Source Term Evaluation Code (ASTEC), to study source
term of a hypothetical severe accident in a nuclear light water reactor
[6-8]. The Japan Atomic Energy Agency (JAEA) has developed an integrated
severe accident analysis code THALES2 and keep extending function via
adding modules like KICHR (Kinetics of Iodine Chemistry in the
Containment of Light Water Reactors) [9, 10]. Nevertheless, most source
term analysis codes focus on water reactor and are not entirely
applicable to HTR.

Source term analysis of HTR-PM has been performed by several
commercial software and empirical formula during the design. All results
adopt conservation estimations and have been appraised by the National
Nuclear Safety Administration (NNSA) of China. At present, commercial
software used to study HTR source term only focuses on the radioactivity
inventory and release of reactor core while source term of the primary
circuit and the release of airborne radioactive materials are not
involved. In order to build a more systematic HTR source term analysis
program package for the following commercial pebble-bed HTR design, a
software package named HTR-STAC (HTR-PM Source Term Analysis Code) is
developed. In this article, some prominent features of HTR-STAC are
described and the code assessment is also given.

2. The Features of HTR

HTR-10 (10 MW High-Temperature Gas-cooled Test Reactor) is the
first experimental high-temperature gas-cooled reactor in China, which
is designed and constructed in the 1990s, brought to criticality in
2000, and reached full power operation in 2003 [11] and commercial 600MW
High-Temperature Reactor-Pebble-bed Modules (HTR-PM600, Figure 2) will
be the next HTR in China (Figure 1). Table 1 shows some main parameters
of HTR-10, HTR-PM, and HTR-PM600. It can be seen that HTR-PM and
HTR-PM600 share most of the characteristic parameters while some HTR-10s
are lower. The lower value is designed to make HTR-10 more controllable
to perform tests and experiments. The "250 x 2" reactor
thermal power of HTR-PM represents its two 250 [MW.sub.th] reactor
modules, and HTR-PM600 has six reactor modules. There is just one steam
turbine in HTR-PM and the same to HTR-PM600. What is more, the three
HTRs have the same coolant, fuel type, and on power refueling model.

HTR-PM is taken, for example, to introduce some most important
features of HTR.

As mentioned above, HTR-PM consists of two pebblebed reactor
modules coupled with a 210 MW steam turbine. Each reactor module
includes a reactor pressure vessel; graphite, carbon, and metallic
reactor internals; a steam generator; and a main helium blower. HTRs use
graphite as a moderator as well as structural material and helium as a
coolant which could reach 750[degrees]C at the core outlet. [12]

Spherical fuel element with a diameter of 60 mm (Figure 3) is used
in HTR-PM. Each fuel element contains about 12,000 coated particles
which are uniformly embedded in a graphite matrix of 50 mm in diameter
and an outer fuel-free zone of pure graphite surrounds the fuel graphite
matrix. A coated fuel particle is composed of an U[O.sub.2] kernel of
0.5 mm diameter and three pyrolytic carbon (PyC) layers and one SiC
layer (TRISO) [13]. Experimental results show that the spherical fuel
element will effectively be retained under 2200[degrees]C, which exceeds
the safety limit of 1620[degrees]C for any operating or accident
condition. The heat-resistant property of spherical fuel element ensures
core would not melt down.

A unique fuel-discharge system allows the operation mode of HTR-PM
to adopt continuous fuel loading and discharging. Fuel elements go
through the core by gravity from up and down and are discharged through
a fuel extraction pipe at the core bottom. The discharged fuel elements
would be measured one by one to check their states of burn-up. A fuel
element will be transported into the spent fuel storage tank if it
reaches the design burn-up; otherwise, it will pass the core once again.

Average core power density of HTR-PM is about 3 MW/[m.sup.3] while
pressurized water reactor about 100 MW/[m.sup.3]. Lower power density
means greater thermal-inertia and a slower rise in core temperature
under accident conditions. Besides, two independent shutdown systems are
installed in HTR-PM: a control rod system and a small absorber sphere
(SAS) system.

HTR-10, HTR-PM, and HTR-PM600 share the same operation model and
physical process, resulting in the same source term generation
mechanism. Hence, HTR-STAC could be applied to the source term analysis
of all these HTRs.

3. HTR-STAC Description

The objective of HTR-STAC is to study the accumulation and release
of several significant nuclides, such as cesium, strontium, silver,
iodide, tritium, etc., in normal and accident condition of pebble-bed
HTR. The structure of HTR-STAC is modular and it consists of five units,
including LOOP (Primary Circuit Source Term Analysis Code), NORMAL
(Normal Condition Airborne Source Term Analysis Code), ARCC (Accident
Release Category Calculation code), CARBON (C-14 Source Term Analysis
Code), and TRUM (Tritium Source Term Analysis Code) (Figure 4). Each
unit can be run independently for separate tests or coupled to take the
overall evaluation.

The programming language is Python and the code runs on a PC in
diverse environments such as Linux and Windows. In addition, a visual
interface with parameters input, results display, and output documents
download function is being developed (Figure 5).

3.1. LOOP. To estimate the effect of the most serious accident,
i.e., the core melt accident, the amount of core radioactivity has been
a significant issue for reactors for a long time. Based on the special
design, the special spherical fuel element will be perfect under any
accidental condition and there is no melt down of the reactor core.
However, fission product (FP) still release from fuel elements would
transport in the primary circuit via helium cycle. The primary
radioactivity could release slowly under norm operating condition and
would be a major source of radioactivity release during an accident.
Hence, the primary circuit coolant radioactivity under normal operating
conditions of HTRs should be studied and LOOP aims to do that.

FPs in helium is mainly generated in two ways: coated fuel
particles failure and uranium contamination. It is found that a very
small amount of TRISO particles with a defect layer in spherical fuel
elements during the fabrication process and the irradiation would also
induce few coated fuel particles failure [18]. Uranium contamination
mainly exists on the surface of the coating layer and matrix graphite
and sometimes also on natural graphite. Furthermore, the continuous
reductions of FPs caused by atom decay, helium purification system, and
deposition on the primary circuit surface should be considered in the
calculation. The dynamic equation of FPs in the primary circuit could be
addressed as

FP i generated by activation of materials inside the primary
circuit are also considered, and the dynamic equation is given by

[B.sub.i] = [[rho][A.sub.O][f.sub.n][f.sub.m]/A] x [sigma][phi] (2)

[B.sub.i]: the generation rate of FP i in per unit volume of
material (m-3 x [s.sup.-1]).

[rho]: material density (g x [cm.sup.-3]).

[A.sub.O]: Avogadro's constant ([mol.sup.-1]).

[f.sub.n]: natural abundance of target nuclide (%).

[f.sub.m]: the weight percentage of the target element in the
material (%).

A: molar mass of target nuclide t (g x [mol.sup.-1]).

[sigma]: neutron absorption cross section of Ar-40 (cm2).

[phi]: neutron fluence rate (cm-2 x [s.sup.-1]).

Based on this equation, LOOP calculates several radioactive
nuclides amount chosen by the user in primary circuit coolant, i.e.,
coolant source term analysis under Normal operating conditions.

3.2. NORMAL. During the operation of pebble-bed HTR, the airborne
radioactive material is considered to be the main source of radioactive
discharge. It is essential to study this issue for safe areas division
and radiation level assessment. Therefore, NORMAL has been developed to
study the release of airborne radioactive materials to the environment
under normal operating conditions.

Six airborne radioactivity sources are considered (taking HTR-PM,
as instance, shown in Figure 6) and calculated individually in NORMAL,
including

(A) air activation inside the cavity;

(B) leakage of primary coolant;

(C) venting of contaminated He tank;

(D) venting of fuel-discharge system;

(E) leakage of secondary loop steam;

(F) leakage of equipment room during maintenance.

A cavity negative pressure air exhausting system has been installed
in HTR-PM to discharge cavity air after filtration. However, Ar-41, a
radionuclide generated by neutron activation from Ar-40, which shares
0.93% of air, cannot be filtrated and would release to atmosphere
through the system at the same time. The dynamic equation is given by

A little helium leaking from primary coolant and some radioactive
gas generated by fuel-discharge system and maintenance of equipment room
also release through negative pressure air exhausting system with
filtration.

Waste helium in contaminated helium tank is mainly produced in the
regeneration of helium purification system. Because of the decline of
purification equipment's ability to transform or adsorb impurities,
helium purification system needs regeneration after ten days. Desorption
of parts of absorbed radioactive nuclides in this process makes them
contribute to airborne radioactivity, too.

Due to the activation and penetration of some radioactive
materials, the secondary loop steam also has certain radionuclides and
they are mainly tritium. The amount of airborne radioactivity leaked
from the secondary loop steam is related to the concentration of tritium
in the secondary loop and the operating status of the turbine.

NORMAL estimates the six parts' air activation and presents an
output result including each part's radioactivity and total
evaluation.

SGTR (Steam Generator Tube Rupture), LOCA (Loss of Coolant
Accident), and Transient Process are three typical accidents on pebbled
bed HTR. They are all considered in ARCC code and could represent most
design basic accidents (DBA) in HTR. However, the source term analysis
of beyond design basic accidents (BDBA) needs to be proceeded by a
different way so they are not included in ARCC.

Generally speaking, there are two release processes in the accident
progress. One is transient release and the other is long-term release.
ARCC firstly classifies accidents by several input accident parameters
and then outputs instant release, long-term release, and a total
release, respectively. Input parameters describe the characteristic and
states of an accident, including fuel temperature, valve state,
filtration efficiency, flooding quantity, etc. The long-term release
comes from radioactive fission products in fuel elements caused by the
heat up of the core after accidents and it is mainly decided by core
temperature. However, instant release varies from accident to accident
and should be studied case by case.

3.3.1. SGTR. Steam generator of HTR-PM has 19 heat exchange pipes
(Figure 7). The pressure of primary circuit of HTR-PM is lower than the
secondary circuit, so once a break occurs, water and water vapour in the
secondary circuit would rapidly flow into the primary circuit and wash
out radioactivity deposited on the steam generator and primary circuit.
A higher pressure would trigger safety valves by which radioactivity
release through. According to the size of the tube break, SGTR falls
into three types:

(A) Small break of the heat pipe, represented by the double-end
fracture of one heat pipe.

(B) Large break of the heat pipe, represented by the double-end
fracture of several heat pipes.

(A) The primary circuit coolant radioactivity during steady
operation before the accident which could be studied by LOOP.

(B) The radioactivity deposited on the inner surface of the steam
generator washed into the primary coolant.

(C) The radioactivity caused by the reacts between water vapour and
matrix graphite or broken fuel element.

3.3.2. LOCA. LOCA of HTRs is the same as water reactors except the
coolant is helium instead of water. The high-pressure helium with
radioactive materials discharged into the containment vessel through a
primary circuit break will cause a pressure increase and then trigger
rupture discs so that airborne radioactivity would release to the
environment.

The accident severity levels of LOCA are decided by the break size
of primary circuit tube. Radioactivity instant release in LOCA also
consists of three parts:

(A) The primary circuit coolant radioactivity during steady
operation before the accident, the same as SGTR.

(B) Desorption of adsorptive radioactivity deposited on the inner
surface of the primary circuit.

3.3.3. Transient Process. The Transient Process is an accident
which causes a turbine trip in addition to SGTR and LOCA. Transient
Process also causes a pressure increase and the radioactivity instant
release in the Transient Process is the same as LOCA, but release
through safety valve instead of break pipe.

3.4. CARBON and TRUM. C-14 and tritium are two special
radionuclides in HTR which needed to be studied separately. C-14 exists
mainly in the form of carbon dioxide in the environment and could be
easily entered into the human body through carbon cycling and a
5730-year half-life makes its influence cannot be ignored. Tritium oxide
(HTO, DTO, or T2O) can be inhaled and can combine with organic matter
and hard to excrete, resulting in internal irradiation, which is very
harmful to the human body, too. C-14 and tritium are generated by two
ways: ternary fission and neutron reactions (Tables 2 and 3). CARBON and
TRUM study the cumulative quantity and cumulative rate of C-14 and
tritium of each reaction in a specified time which can be set by users.

Neutron activation reaction should be calculated by corresponding
formula while the amount of H-3 or C-14 generated per unit of time by
ternary fission reaction of heavy nuclides can be given by

d[N.sup.X.sub.T]/dt = [P/g] x y(X) x f(X) (4)

X: H-3 or C-14.

P: reactor power (MW).

f: the fission reaction share of nuclide X (%).

y: the yield of light nuclide (%).

g: the energy released from each fission reaction (3.27 x
10-17MW-s/each fission).

4. HTR-STAC Code Assessment

According to IEEE-1012 and HAF-102 (Quality Assurance for Safety in
Nuclear Plants), we conducted the verification and validation of
HTR-STAC.

Beijing QunYuan Power Technology Co., Ltd., helped us perform the
first verification. They finished the code reading and algorithm
analysis and performed several examples to confirm that HTR-STAC could
study HTR source term.

We also submitted the code, documentation, and development report
to NNSA and they performed the verification of HTR-STAC with the safety
analysis report of HTR-PM. NNSA performed static analysis on concept
verification, requirements verification, and design verification of
HTR-STAC and they performed dynamic analysis on implementation
verification and installation and checkout verification of HTR-STAC.
They also finished software requirements evaluation, criticality
analysis, configuration management assessment, hazard analysis, security
analysis, and risk analysis. NNSA confirmed HTR-STAC satisfied the
requirements of HTR source term analysis.

And all the subprograms of HTR-STAC have been conducted code
validation in combination with other researches or the real operation
data. HTR-10 is chosen to obtain operation data. The Institute of
Nuclear and New Energy Technology (INET), Tsinghua University, has
performed several data-collection campaigns on HTR-10, such as source
term experimental analysis of irradiated graphite in the core, sampling
of the radioactive graphite dust in the primary loop, and R&D of
helium sampling loop [19-22]. INET also has studied many theoretical
calculation approaches like the Monte Carlo method and differential
equations to predict the radioactivity in HTR-10 [14, 17, 23, 24].

Validation examples of five modules are described below.

4.1. LOOP. This example is based on real data of HTR-10 (Figure 8).
The experiment results, theoretical calculation results, and LOOP
running results of radioactive activity of several nuclides in primary
circle coolant are showed in Table 4.

The LOOP set temperature at 475[degrees]C, time of He cycle is
3.13s, and purification flow is 6100 c[m.sup.3]/s. The uncertainty of
experiment consists of 4 part [14]:

(a) The uncertainty of the instrument count, no more than 5%.

(b) The uncertainty of the counter detection efficiency, no more
than 10%.

(c) The uncertainty of delay time between sampling and measurement,
no more than 12%.

(d) Other uncertainty, no more than 3%.

It is shown that the activity of Kr-85m and Kr-88 of LOOP is nearly
the same to experiment; the activity of Kr-87 and Xe-133 of LOOP is
closer to experiment than theoretical results; the running results of
Xe-135 and Xe-135m are not as good as theoretical calculation results to
estimate real situation. Due to the fact that reference article has not
provided detailed experiment parameters like reactor operating time and
coolant temperature, the LOOP results, which are affected by input
parameters, may deviate from experimental data. In general, running
results of LOOP have good reference value.

4.2. NORMAL. This example is the comparison of design value and
NORMAL calculation results of some typical airborne radioactive
discharge of HTR-PM under normal operation condition (Table 5). It is
shown that the calculation results of H-3, C-14, and iodine isotopes
agree well with design value while the difference of Ar-41 may be due to
the different estimation of the cavity space.

4.3. ARCC. Due to the safely running of HTR-10, accident
experimental data cannot be obtained. Hence, we compared the ARCC
results with YuanZhong Liu's research [16] and take a LOCA accident
as an example.

It supposes that a 65 mm diameter pipe break causes the accident,
and then the airborne radioactivity releases to the environment. The
release of several typical nuclides is shown in Table 6.

It can be seen that ARCC's results are lower that may be due
to YuanZhong Liu made some more conservative assumption, such as a
higher deposition rate, more graphite dust in He, to simplify model.
However, two results are still in the same magnitude, showing ARCC agree
with Yuan-Zhong Liu's research.

4.4. CARBON. YuanZhong Liu's research also shows the
radioactivity of C-14 in the primary loop and some necessary reactor
parameters. The operational thermal power of the reactor is 10 MW,
primary helium pressure is 3 MPa, average burn-up is 80 GWd/tU, and fuel
elements are 2700. The corresponding CARBPON input parameters are same
with above values and other input parameters take the most conservative
value. AS it is shown in Table 7, CARBON result is about 1.5 times
YuanZhong Liu's due to the most conservative assumption.

4.5. TRUM. The comparison of experiment result, theoretical
calculation, and TRUM running result of tritium in the primary loop of
HTR-10 are showed in Table 8. Most input parameters of TRUM are same as
theoretical calculation and as close as possible to the experimental
conditions. The operational thermal power of the reactor is 2.9 MW,
primary helium pressure is 2.1 MPa, primary helium flow rate is 1.39
kg/h, and total He inventory in the primary circuit is [1200.sub.STP]
[m.sup.3].

The theoretical calculation result is about 20 times higher than
experiment data, which shows conservatism for the evaluation of the
activity concentration of tritium in the primary loop of HTR-10. The
calculation result of HTR-STAC is also conservative and closer to actual
data, which indicates that HTR-STAC is valid and more adoptive than
previous theoretical calculation. The discrepancy between HTR-STAC data
and experimental data may come from the most conservative input
parameters, especially the concentration of He-3 in the helium coolant
and Li-6 in the graphite.

5. Conclusion and Remarks

HTR-STAC is a compositive Source Term Analysis Code for Pebble-bed
HTR, consisting of LOOP, NORMAL, ARCC, CARBON, and TRUM. Each subroutine
of HTR-STAC could run independently or unites together. All subprograms
are performed code validation and the results show that HTR-STAC
analysis is closer to experiment than previous theoretical calculation.

Source term analysis of HTR-PM has done before the development of
HTR-STAC and the results are reliable. Many algorithms, empirical
formulas, and assumptions during the design and safety analysis of
HTR-PM are also applied in HTR-STAC. HTRs share the same operation model
and physical process so that HTR-STAC could systematically perform
source term analysis for most HTRs conveniently and flexibly. In the
near future, more pebble-bed HTRs will be design and HTR-STAC will make
source term analysis of HTRs more reliable and accurate and would make a
great contribution to the design and promotion of pebble-bed HTR.

https://doi.org/10.1155/2018/7389121

Data Availability

The data used to support the findings of this study are available
from the corresponding author upon request.

Conflicts of Interest

The authors declare that they have no conflicts of interest.

Acknowledgments

This work has been supported by the National S&T Major Project
(Grant no. ZX069).